To non-destructively determine the burnup of a spent nuclear fuel assembly, it is essential to analyze the nuclear isotopes present in the assembly and detect the neutrons and gamma rays emitted from these isotopes. Specifically, gamma-ray measurement methods can utilize a single radiation measurement value of 137Cs or measure based on the energy peak ratio of Cs isotopes such as 134Cs/137Cs and 154Eu/137Cs. In this study, we validated the extent to which the results of gamma-ray measurements using cadmium zinc telluride (CZT) sensors based on 137Cs could be accurately simulated by implementing identical conditions on MCNP. To simulate measurement scenarios using a lead collimator, we propose equations that represent radiation behavior that reaches the detector by assuming “Direct hit” and “Penetration with attenuation” situations. The results obtained from MCNP confirmed an increase in measurement efficiency by 0.47 times when using the CZT detector, demonstrating the efficacy of the measurement system.

Concerning the apprehensions about naturally occurring radioactive materials (NORM) residues, the International Atomic Energy Agency (IAEA) and its member nations have acknowledged the imperative to ensure the radiation safety of NORM industries. Residues with elevated radioactivity concentrations are predominantly produced during NORM processing, in the form of scale and sludge, referred to as technically enhanced NORM (TENORM). Substantial quantities of TENORM residues have been released externally due to the dismantling of NORM processing factories. These residues become concentrated and fixed in scale inside scrap pipes. To assess the radioactivity of scales in pipes of various shapes, a Monte Carlo simulation was employed to determine dose rates corresponding to the action level in TENORM regulations for different pipe diameters and thicknesses. Onsite gamma spectrometry was conducted on a scrap iron pipe from the titanium dioxide manufacturing factory. The measured dose rate on the pipe enabled the estimation of NORM concentration in the pipe scale onsite. The derived action level in dose rate can be applied in the NORM regulation procedure for on-site judgments.

The nuclear fuel that melted during the Fukushima nuclear accident in 2011 is still being cooled by water. In this process, contaminated water containing radioactive substances such as cesium and strontium is generated. The total amount of radioactive pollutants released by the natural environment due to the nuclear accident in Fukushima in 2011 is estimated to be 900 PBq, of which 10 to 37 PBq for cesium. Radioactive cesium (137Cs) is a potassium analog that exists in the water in the form of cations with similar daytime behavior and a small hydration radius and is recognized as a radioactive nuclide that has the greatest impact on the environment due to its long half-life (about 30 years), high solubility and diffusion coefficient, and gamma-ray emission. In this study, alginate beads were designed using Prussian blue, known as a material that selectively adsorbs cesium for removal and detection of cesium. To confirm the adsorption performance of the produced Prussian blue, immersion experiments were conducted using Cs standard solution, and MCNP simulations were performed by modeling 1L reservoir to conduct experiments using radioactive Cs in the future. An adsorption experiment was conducted with water containing standard cesium solution using alginate beads impregnated with Prussian blue. The adsorption experiment tested how much cesium of the same concentration was adsorbed over time. As a result, it was found that Prussian blue beads removed about 80% of cesium within 10-15 minutes. In addition, MCNP simulation was performed using a 1 L reservoir and a 3inch NaI detector to optimize the amount of Prussian blue. The results of comparing the efficiency according to the Prussian volume was shown. It showed that our designed system holds great promise for the cleanup and detection of radioactive cesium contaminated seawater around nuclear plants and/or after nuclear accidents. Thus, this work is expected to provide insights into the fundamental MCNP simulation based optimization of Prussian blue for cesium removal and this work based MCNP simulation will pave the way for various practical applications.

For efficient design and manufacture of PWR spent fuel burnup detector, data simulated with various condition of spent fuel in the NPP storage pool is required. In this paper, to derive performance requirements of spent fuel burnup detector for neutron flux and dose rates were evaluated at various distances from CE16 and WH17 types of fuel, representatively. The evaluation was performed by the following steps. First, the specifications of the spent fuel, such as enrichment, burnup, cooling time, and fuel type, were analyzed to find the conditions that emit maximum radioactivity. Second, gamma and neutron source terms of spent fuel were analyzed. The gamma source terms by actinides and fission products and neutron source terms by spontaneous and (α, n) reactions were calculated by SCALE6 ORIGAMI module. Third, simulation input data and model were applied to the evaluation. The material composition and dose conversion factor were referred as PNNL-15870 and ICRP-74 data, respectively and dose rates were displayed with the MCNP output data. It was assumed that there was only one fuel modeled by MCNP 6.2 code in pool. The evaluation positions for each distance were selected as 5 cm, 10 cm, 25 cm, 50 cm, and 1 m apart from the side of fuel, respectively. Fourth, neutron flux and dose rates were evaluated at distance from each fuel type by MCNP 6.2 code. For WH 17 types with a 50 GWd/MTU burnup from 5 cm distance close to fuel, the maximum neutron flux, gamma dose rates and neutron dose rates are evaluated as 1.01×105 neutrons/sec, 1.41×105 mSv/hr and 1.61×101 mSv/hr, respectively. The flux and dose rate of WH type were evaluated to be larger than those of CE type by difference in number of fuel rods. The relative error for result was less than 3~7% on average secured the reliability. It is expected that the simulated data in this paper could contribute to accumulate the basic data required to derive performance requirements of spent fuel burnup detector.

Some of the metal waste generated from KEPCO NF is being disposed of in the form of ingots. An ingot is a metal that is melted once and then poured into a mold to harden, and it is characterized by a uniform distribution of radioactive material. When measuring the uranium radioactivity in metal ingot with HPGe detector, 185.7 keV of U-235 is used typically because most gamma rays emitted at U-235 are distributed in low-energy regions below 200 keV. To analyze radioactivity concentration of U-235 with HPGe detector more accurately, self-attenuation due to geometrical differences between the calibration source and the sample must be corrected. In this study, the MCNP code was used to simulate the HPGe gamma spectroscopy system, and various processes were performed to prove the correlation with the actual values. First an metal ingottype standard source was manufactured for efficiency calibration, and the GEB coefficient was derived using Origin program. And through the comparison of actual measurements and simulations, the thickness of the detector’s dead layers were defined in all directions of Ge crystal. Additionally instead of making an metal ingot-type standard source every time, we analyzed the measurement tendency between commercially available HPGe calibration source (Marinelli beaker type) and the sample (metal ingot type), and derived the correction factor for geometry differences. Lastly the correction factor was taken into consideration when obtaining the uranium radioactivity concentration in the metal ingot with HPGe gamma spectroscopy. In conclusion, the U-235 radioactivity in metal ingot was underestimated about 25% of content due to the self-attenuation. Therefore it is reasonable to reflect this correction factor in the calculation of U-235 radioactivity concentration.

Recently, the spent fuel pools withdrawn from nuclear power plants in Korea have been saturated. Therefore, specific regulations on the management of spent fuel pools, such as transportation and intermediate storage are needed. The burnup history is directly related to the management of spent nuclear fuel. This is because the decision to handle nuclear fuel may vary depending on the initial concentration of nuclear fuel, the degree to which nuclear fuel is irradiated and radioisotope nuclides are decayed, and the cooling state in the spent nuclear fuel storage tank. The purpose of this study is to determine the burnup of fuel based on the value obtained by scanning the surface of spent nuclear fuel through a neutron detector. Conversely, a database of neutron signals that scan bundles of spent nuclear fuel with an instrument with an already identified combustion history needs to be completed. First of all, the correlation between burnup history and nuclides was identified in previous studies. By setting the burnup history as the input value in the ORIGEN-ARP code, it was possible to identify the radioactive isotopes remaining in the bundle of nuclear fuel. Neutrons can finally be measured based on the amount of nuclide inventory that constitutes spent nuclear fuel. Through MCNP, the neutron detector was simulated and signals were measured to confirm how it correlates with the previously acquired burnup history database. In addition, the M (sub-critical multiplication) value, which is essential for neutron measurement, was checked to confirm the degree to which additional neutrons were generated in spent nuclear fuel in a subcritical state. The target nuclear fuel assembly was CE16×16, WH14×14, and WH17×17, which confirmed the correlation (1) between burnup, enrichment, and cooling time with the previous research topic, TNSI (Total neutron source intensity). 􀜤􀜷􁈺􀜩􀜹􀝀/􀜯􀜶􀜷􁈻 = 0.83􁈺􀜵􀯇􁈻􀬴.􀬶􀬷􀬼 ∙ 􁈺􀜫􀜧􁈻􀬴.􀬸􀬺􀬶􀬻 ∙ 􀝁􀬴.􀬴􀬴􀬼􀬷∙􀯧 􁈺1􁈻 A neutron signal will be obtained from the case according to each burnup history constituting this database. In particular, PAR=SF, a function that calculates the production amount of the fission product, was used. To confirm the computational logic of SF, it was confirmed whether a reasonable calculation was made by calculating with a nuclide spectrum.

Medical cyclotrons have been used for dedicated medical of commercial applications such as positron emission tomography (PET) for the past tens of years. These cyclotron facilities have produced positron-emitting radionuclides (i.e. 11C, 13N, 15O, 18F, etc.). Among them, 18F, produced by 18O(p,n)18F reaction is the most widely used which has longer half-life (around 110 m) and lower energy of emitted positrons (around 0.63 MeV). Secondary neutrons produced during 18O(p,n)18F reaction could cause neutron activation of structures, systems, and components of cyclotron facilities. Therefore, International Atomic Energy Agency (IAEA) had addressed that during the operation of cyclotrons, concrete walls become radioactive over time and this radioactivity needs to be characterized for planning of the facility decommissioning. Moreover, several prior studies had estimated the neutron activation and levels of radioactivity of concrete wall of cyclotron facilities. Although those studies assessed the neutron activation of actual cyclotron facilities, however, the purpose of assessment was only for decommissioning each individual facility. Also, the assumptions, conditions or insights of conclusion may be limited to each individual case. For these reasons, this study focused on analysis of effects of major factors (e.g. concrete type, impurity contents of structural materials, etc.) about neutron activation of cyclotron facilities. In this study, the well-known methodology of neutron activation estimation was established and neutron activation products of concrete wall of cyclotron vault was calculated. Also, sensitivity analyses were conducted to figure out the effects of major factors of neutron activation and production of radioactive wastes during decommissioning of the facility. The methodology and results were validated by two steps: comparing with prior studies and comparing with another computer code. Concrete type did not affect that the decision of level of radioactivity waste criteria. Because of relatively longer half-lives, impurity contents of structural materials especially Co and Eu were turned out one of the most important factors for planning the facility decommissioning. It is hard to simply figure out the radioactivity levels of cyclotron facilities, however, rough predictions of minimum period for decay-in-storage as radioactive waste management can be possible with using information of thermal neutron spectra and major impurity nuclides (e.g. 59Co, 151Eu and 153Eu) for minimization of radioactive waste production and relief of charge of radioactive waste management.

Important medical radionuclides for Positron Emission Tomography (PET) are producing using cyclotrons. There are about 1,200 PET cyclotrons operated in 95 countries based upon IAEA database (2020). Besides, including PET cyclotrons, demands for particle accelerators are continuously increasing. In Korea, about 40 PET cyclotrons are in operating phases (2020). Considering design lifetime (about 30-40 years) and actual operating duration (about 20-30 years) of cyclotrons, there will be demands for decommissioning cyclotron facilities in the near future. PET cyclotron produces radionuclides by irradiating accelerated charged particles to the targets. During this phase, nuclear reactions (18O(p,n)18F etc.) produce secondary neutrons which induce neutron activation of accelerator itself as well as surrounding infrastructures (the ancillary subsystems, peripheral equipment, concrete walls etc.). Generally, experienced cyclotron personnel prefer an unshielded cyclotron because of the repair and maintenance time. In unshielded cyclotron, water cooling systems, air compressor, and other equipment and structures could be existed for operating purposes. Almost all the equipment and structures are consisted of steel, and these affect neutron distribution in vault especially thermal neutron on the concrete wall. In addition, most of them can be classified as very low level radioactive wastes by Nuclear Safety and Security notice (NSSC Notice No. 2020-6). However, few studies were estimating radioactivity concentrations (Bq/g) of surrounding structures using mathematical calculation/simulation codes, and they were not evaluating the effect of surrounding structures on neutron distribution. In this study, by using computational neutron transport code (MCNP 6.2), and source term calculation code (FISPACT- II), we evaluated effect of the interaction between surrounding structures (including surrounding equipment) and secondary neutrons. Discrepancies of activation distribution on/in concrete wall will be occur depending on thickness of structure, distance between structures and walls, and consideration of interaction between structures and neutrons. Throughout this study, we could find that the influence of those structures can affect neutron distribution in concrete walls even if, thickness of the structure was small. For estimating activation distribution in unshielded cyclotron vault more precisely, not only considering cyclotron components and geometry of target, but also, considering surrounding structures will be much more helpful.

Korea Institute of Radiological and Medical Sciences provides proton irradiation service of up to 40 MeV using cyclotron. The use of such a cyclotron was approved in advance to satisfy the Nuclear Safety Act, and radiation safety was evaluated in this process. The Monte Carlo method is generally used to evaluate the shielding safety of high-energy accelerators, and MCNP 6.2 was used in the previous evaluation. In this study, in order to verify the results of previous evaluation, the calculation results of MCNP 6.2 and Particle and Heavy Ion Transport code System (PHITS) 3.24 are compared. PHITS is a general-purpose Monte Carlo particle transport simulation code that is used in many studies in the fields of accelerator technology, radiotherapy, space radiation, etc. In the previous evaluation, the effective dose by neutrons and photons generated by the collision of 40 MeV 20 μA of protons with a 10.5 mm thick beryllium target was evaluated, and in this study, this was reproduced with PHITS. As the radiation exposure evaluation for the user or pubic is evaluated based on the radiation dose and energy distribution generated around the target, the effective dose and energy distribution received by the water phantom with a radius of 1 cm on the front, side, and back of the target were calculated. T-Track, a tally of PHITS, was used to calculate effective dose, which is similar to F4 tally of MCNP 6.2 using a dose conversion factor. For the dose conversion factor, the value suggested as AP irradiation in Publication 103 was used. As a result of the calculation, the effective dose by neutrons at the front, side and back of the target was 1.42×105, 2.09×104, and 1.39×104 mSv·h−1, respectively, which was similar to 2.00×105, 1.84×104, and 2.59×104 calculated using F4 tally in MCNP. Moreover, the results of calculating the effective dose by photons using PHITS were 4.81×10, 3.10×10, and 2.66×10, respectively, and the results of calculating MCNP were 4.49×102, 6.45×10, and 9.64×10. The average energies of neutrons were 11.2, 0.69, and 0.31 MeV when calculated by PHITS, respectively, and 13.8, 7.8, and 4.6 when calculated by MCNP. Moreover, the average energies of photons were 1.98, 0.98, and 0.86 when calculated by PHITS, respectively, and 3.9, 3.2, and 2.6 when calculated by MCNP.

RADTRAN is a code that assesses the radiation risk of radioactive material transportation. RADTRAN assumes that the package is a point source or a line source regardless of package type and corrects the external dose rate using a shape factor which depends on the critical dimension of the package. However, the external dose rate calculated using a shape factor may be different from the actual external dose rate. Therefore, it is necessary to analyze the effect of the shape factor on the external dose rate. In this study, the effect of the shape factor on the external dose rate in RADTRAN was analyzed by comparison with MCNP. This study analyzed change in external dose rate depending on the distance from the package and the critical dimension. The distance from the package was in the range of 1–800 m. The shape of the package was assumed to be cylindrical with a radius of 1 m, and the critical dimensions of the package were assumed to be 2, 4, and 8 m. Attenuation and build-up in the air were not considered to consider only the effect on the shape factor. When simulating the exposure situation using MCNP, the package was assumed to be a volume source, and flux by distance from the package was calculated using F5 tally. The dose rate at 1 m from the package was normalized to 2 mSv·hr−1. As a result of the analysis, the external dose rates of the package were higher in RADTRAN than in MCNP. For the critical dimension of 2, 4, and 8 m, when the distance from package is 1–10 m, the RADTRAN was 1.83, 4.08, and 5.27 times higher on average than MCNP, respectively. And when the distance from the package was 10–100 m and 100–800 m, RADTRAN was 1.10, 2.02, 3.01 times and 1.04, 1.92, 2.43 times higher than MCNP, respectively. It was found that the larger the distance from the package is and the smaller the critical dimension of the package is, the less conservatively RADTRAN assessed. It is because the shape of the package gets closer to the point source as the distance from the package increases, and the shape factor decreases as the critical dimension of the package decreases. The result of this study can be used as the basis for radiation risk assessment when transporting radioactive materials.

To obtain the gamma-ray energy spectrum of artificial radionuclides which is difficult to obtain practically, virtual gamma-ray energy spectrum simulator program was developed. It can be applied for the predetermined measurement condition for which the database was developed through computational simulation and actual measurement of background radiation. For gamma spectrometry training for KHNP HPGe detectors using this program, the database for KNPG HPGe detectors was developed. First, the geometry of the detector in the simulation was adjusted to resemble the real structure by comparing the actually measured net counts rate at the main gamma peak with the value simulated by MCNP6. The Certified Reference material (CRM) of 137Cs and 60Co were used for verification. The comparison was made with respect to the situation where CRM was attached to the top and side of the detection part of the considered detector. The geometry structures of detectors were simulated by reflecting the design drawing of the products, and the simulation was performed for several thicknesses of the Ge/Li dead layer in consideration of the change in the thickness over time. As the results, the simulation geometry was tuned so that the results for 137Cs showed a difference within 10% for all detectors. At this time, in some detectors, the result for 60Co shows a 10% higher error, which is estimated to be due to the random summing. It was not considered in tuning the simulation geometry, but it was found that improvements were needed to reflect the coincidence summing when construction the virtual spectrum in the future. The determined simulation geometry was applied to generate theoretical gamma-ray energy spectra of representative artificial radionuclides. In order to create a virtual spectrum similar to the real one, the background spectrum was measured for each detector without a source, and the simulation results were calculated in the form of having the same energy channel as the background spectrum. The background spectrum and theoretical spectra of artificial radionuclides for each detector were databased so that virtual spectra could be generated under desired conditions. The virtual spectrum was generated by adding a background spectrum and a spectrum obtained by multiplying the spectrum of the desired nuclide by the concentration of the nuclide. The validity of generated virtual spectra was verified using the pre-developed gamma spectrometry program. As a results of gamma spectrometry of virtual spectra, the virtual spectra was verified by showing a difference within 20% from the radioactivity value input when generating the virtual spectra.

For producing radionuclides which were mostly used in medical purposes, for instance, Positron Emission Tomography (PET), there were about 1,200 PET cyclotrons operated in 95 countries based upon IAEA database (2020). Besides, including PET cyclotrons, demands for particle accelerators are continuously increasing. In Korea, about 40 PET cyclotrons are in operating phases (2020). Considering design lifetime (about 30–40 years) of cyclotrons, there will be demands for decommissioning cyclotron facilities in the near future. PET cyclotron produces radionuclides by irradiating charged particles to the targets. During this phase, nuclear reactions (18O(p,n)18F, 14N(d,n)15O etc.) produce secondary neutrons which induce neutron activation of accelerator itself as well as surrounding infrastructures (the ancillary subsystems, peripheral equipment, concrete walls etc.). Most of the ancillary systems including peripheral equipment can be neutron activated, since, most of them were made of steels. Steels like stainless steel or carbon steel may contain some impurities, typically cobalt. Although, there were several researches evaluating activation of concrete walls and accelerator components, estimating the activation and influence on neutron interaction of the other surrounding infrastructures were insufficient. In this study, by using computational neutron transport code (MCNP 6.2), and source term calculation code (FISPACT- II), we estimated neutron distribution in cyclotron vault and activation of ancillary subsystems including some peripheral equipment. Also, using Au foil and Cd cover, we measured thermal neutron distribution at 16 points on the concrete wall, and compared it to calculated results (MCNP). Even though, the compared results matches well, there was a discrepancy of neutron distributions between presence and absence of those equipment. Additionally, in estimating activation distributions by calculating, most of the steel-based subsystems including peripheral equipment should be managed by radioactive wastes after 20 years of operation. Throughout this study, we could find that influence on neutron interaction of those equipment can affect neutron distribution in concrete walls. This results vary the activation depth as well as location of the hot contaminated spot in concrete wall. For estimating or evaluating activation distributions in cyclotron facilities, there was need to consider some equipment located in cyclotron vault.

In worldwide, tens of thousands of units of particle accelerators have been used and more than 97% of those accelerators are used for dedicated medical of commercial applications. Radionuclide production cyclotron produce several positron-emitting radionuclides such as 18F by 18O(p,n)18F reaction which generates secondary neutrons. It is of note that these neutrons cause neutron activation in structures and components of cyclotron facilities. Therefore, International Atomic Energy Agency had addressed that a well-developed estimate of the neutron activation induced radioactive inventory of accelerator facilities is needed for the proper planning and safe implementation of decommissioning using proven methods or codes that can be used to perform activation calculations. Moreover, IAEA suggested that during the operation of cyclotrons, concrete walls become radioactive over time and this radioactivity needs to be fully characterized as part of early decommissioning planning. In this study, Neutron activation in the medical cyclotron facilities was evaluated with the MCNP and FISPACT-II code to analyze the generation of decommissioning radioactive wastes during facilities dismantling. For the reference case, residual radioactivity concentration of each activation product (e.g. 60Co, 152Eu, etc.) was calculated and the sum of fractions of the activity concentration of each radionuclide divided by its clearance level was exceeded 1.0 at each calculation point which means radioactive waste generations during decommissioning of the facility. Several points show the calculated sum of fractions (SoF) at inside wall were bigger than the surface wall. The reason of these phenomena is that the slowdown of the incident neutron energy at the inside wall due to neutron attenuation and larger thermal neutron flux than surface wall. It is of note that each activation reaction cross-section was dominant at thermal neutron energy band. Sensitivity analysis was conducted to analyze the effects of design characteristics (e.g. beam energy and current, operation period, and workload). The SoF was exceeded 1.0 at the least activation condition (i.e. 9 MeV, 10 μA) if the operation period was 10 years. For the realistic condition such as 13 MeV, only 10 μA of beam current case shows the SoF was under union. On the other hand, 19 MeV, 60 μA, and 10 years operation case shows the SoF as 20.4 which means the clearance rule can be applied only after 21 years of decay-in-storage. The result of this study can be used for proper planning of decommissioning and/or new installation of cyclotron facilities include considerations of radioactive waste management.

North Korea claimed to have tested a hydrogen bomb in its fourth nuclear test in 2016, and declared that the hydrogen bomb was completed after the sixth nuclear test in 2017. North Korea’s operation of Yongbyon Graphite-moderated reactor has been thought to be aimed at producing plutonium, but it has been strongly argued that the restart of the Graphite-moderated reactor is, indeed, aimed at supplying tritium recently. Tritium can be used not only to manufacture hydrogen bombs, but also to miniaturize nuclear weapons, making it as a key material for nuclear weapon capability. Since upgrading nuclear weapons and developing hydrogen bombs through the use of tritium by North Korea could pose a major threat to the security of the Korean Peninsula, it is important to accurately evaluate North Korea’s nuclear weapon capabilities through the analysis of nuclear material production scenarios based on its nuclear facilities. However, researches on North Korea’s nuclear materials such as HEU (Highly Enriched Uranium) and Pu production has been actively conducted, while no research has been shown on tritium production yet. Therefore, this study aims to evaluate the tritium productivity based on the analysis of hypothetical nuclear material production facilities and possible tritium production scenarios. Basic research was conducted about the existing theoretical methodology for tritium production, the analysis of the global tritium production history, and the analysis of nuclear facilities. Based on this basic investigation, feasible tritium production scenarios were constructed. Subsequently, based on design criteria of an hypothetical Graphite-moderated reactor, possible tritium production scenario was modeled by applying the TPBAR (Tritium Production Burnable Absorber Rod). In addition, the factors such as 6Li concentration, design parameters, material compositions, and the number of TPBARs, which may affect tritium throughput were analyzed in terms of sensitivity study such that the maximum and minimum throughput can be predicted.

When a radiation detector is applied to the measurement of the radioactivity of high-level of radioactive materials or the rapid response to the nuclear accident, several collimators with the different inner radii should be prepared according to the level of dose rate. This makes the in-situ measurement impractical, because of the heavy weight of the collimator. In this study, an IRIS collimator was developed so as to have a function of controlling the inner radius, with the same method used in optical camera, to vary the attenuation ratio of radiation. The shutter was made to have the double tungsten layers with different phase angles to prevent the radiation from penetrating owing to the mechanical tolerance. The performance evaluation through the MCNP code was conducted by calculating the attenuation ratio according to the inner radius of the collimator. The attenuation ratio was marked on the outer scale ring of the collimator. It is expected that when a radiation detector with the IRIS collimator is used for the in-situ measurement, it can change the attenuation ratio of the incident photon to the detector without replacing the collimator.

Source term analysis should be carried out to prepare the decommissioning of the nuclear power plant. In the planning phase of decommissioning, the classification of decommissioning wastes and the cost evaluation are performed based on the results of source term analysis. In this study, the verification of MCNP/ORIGEN-2 model is carried out for preliminary source term calculation for Wolsung Unit 1. The inventories of actinide nuclides and fission products in fuel bundles with different burn-up were obtained by the depletion calculation of MCNPX code modelling the single channel. Two factors affecting the accuracy of source terms were investigated. First, the neutron spectrum effect on neutron induced activation calculation was reflected in one-group microscopic cross-sections of relevant radio-isotopes using the results of MCNP simulation, and the activation source terms calculated by ORIGEN-2 using the neutron spectrum corrected library were compared with the results of the original ORIGEN-2 library (CANDUNAU.LIB) in ORIGEN-2 code package. Second, operation history effect on activation calculation was also investigated. The source terms on both pressure tubes and calandria tubes replaced in 2010 and calandria tank were evaluated using MCNP/ORIGEN-2 with the neutron spectrum corrected library if the decommissioning wastes can be classified as a low level waste.